US3993579A - Method of encapsulating solid radioactive waste material for storage - Google Patents
Method of encapsulating solid radioactive waste material for storage Download PDFInfo
- Publication number
- US3993579A US3993579A US05/624,805 US62480575A US3993579A US 3993579 A US3993579 A US 3993579A US 62480575 A US62480575 A US 62480575A US 3993579 A US3993579 A US 3993579A
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- US
- United States
- Prior art keywords
- shapes
- resinous
- vitreous carbon
- storage
- waste material
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- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
- G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
- G21F9/28—Treating solids
- G21F9/30—Processing
- G21F9/301—Processing by fixation in stable solid media
- G21F9/302—Processing by fixation in stable solid media in an inorganic matrix
Definitions
- This invention relates to a method for the preparation of radioactive wastes for storage. More specifically, this invention relates to an improvement in the method for the preparation of high-level radioactive waste solids for storage by encapsulation of the wastes in vitreous carbon.
- the high-level waste solids prepared by the above methods must still be stored for an extended period of time.
- the waste might be stored as granular material in metal canisters either in the open, or in water-filled storage tanks.
- the wastes might be mixed with concrete, asphalt or formed into any one of a number of different mixes of glasses.
- most of these methods have problems attendant with their use for the long-term storage of high-level radioactive wastes.
- metal canisters can leak, releasing radioactivity to the atmosphere, wastes tied up in concrete or in many glasses may be leached, i.e., water may dissolve soluble radioactive salts which could then contaminate the environment. Glasses may also devitrify which will increase leach rates above acceptable levels and both concrete and glass have poor thermal conductivity which may prevent effective dissipation of the thermal energy generated by the radioactive material.
- Additional physical protection and improved thermal conductivity can be provided the vitreous carbon shape by placing a plurality of the shapes into a metal canister in a regular spaced array and filling the canister including the interstices between the shapes with a low-melting molten matrix material such as aluminum.
- the method of the invention has several advantages over the prior art methods for preparing highly radioactive waste solids for storage.
- the resin and radioactive material can be mixed and the shapes formed at room temperature.
- the resistance of vitreous carbon to leaching is much better than concrete and generally superior to the glasses, particularly when the glasses are subject to devitrification.
- Thermal conductivity of the carbon and metal matrix is much better than either concrete, asphalt or glass. In the event that recovery of the wastes is desirable in the future, simple combustion in air of the vitreous carbon will permit easy retrieval.
- the highly radioactive waste material should be a fine powder no more than about 325 mesh in size for best results. Since there is a large decrease in resin volume during the carbonizing process, larger particle sizes may result in the formation of cracks in the vitreous carbon shapes.
- Any solid waste material may be used with the method of the invention as long as it is reasonably inert so that it does not react with the catalyst in the resin. Thus it is preferred that the waste material have been calcined to ensure that only inert oxides are present.
- the amount of waste material in the resinous mixture is generally limited to a mixture consistency which can be easily worked. If shapes, such as rods, which can be formed by extrusion are to be prepared, the resinous mixture may contain up to about 70 weight percent waste material, while if flat sheets or spheroids are to be formed, up to about 50 weight percent waste material is the practical limit.
- the curable resinous material may be any thermosetting resin which is liquid at room temperature and which has a high carbon yield.
- One such resin which was found to give good results was polyfurfuryl alcohol although other thermosetting resins such as phenol-formaldehyde can be used in the practice of this invention.
- One limitation of the method of this invention is that in order to prevent the formation of cracks in the shapes during carbonization the shapes are limited to about 1/4 inch in thickness.
- Three shapes have been found useful with the method of this invention.
- Resinous mixtures containing up to 70 weight percent solid material may be extruded into rods up to 1/4 inch in cross section for curing, carbonization and subsequent storage.
- Mixtures containing up to 50 weight percent may be either cast into sheets up to about 1/8 inch in thickness or formed into small spheriods.
- the spheroids, 200-300 ⁇ m in diameter can be readily prepared by injecting the resinous mixture dropwise into agitated vegetable oil at 100° C. This method is advantageous in that the spheroids are formed and cured within seconds, and could be carbonized in hours.
- Carbonization of the cured resinous material containing the solid radioactive waste is readily accomplished by heating the shapes under a vacuum or in an inert atmosphere to from 600° to 1000° C. at a heating rate of about 6° C. per hour to ensure diffusion of gases and complete carbonization of the shapes. Heating should continue for a period of time sufficient to ensure that total carbonization has taken place throughout the entire shape. For rods and sheets of material, carbonization may require from about 100 to 150 hours, while the spheroids 200 to 300 microns in diameter can be carbonized in 9 to 10 hours.
- vitreous carbon Since the vitreous carbon is brittle and prone to breakage, additional long-term storage protection can be provided by placing a large number of the shapes in a metal canister in a regular spaced array and filling the canister including the interstices between the shapes, with a molten low-melting-temperature matrix material such as aluminum.
- This matrix material would provide additional impact protection for the vitreous carbon, reduce further any possibility of leaching the radioactive material and further improve the thermal conductivity in the waste container.
- An advantage of this method is that recovery of the radioactive material still remains a relatively simple procedure, since only the metal need be melted to recover the vitreous carbon sheets.
- Quaker Oats RP 100A resin polyfurfuryl alcohol
- the batch was catalyzed with 8 w/o Quaker Oats RP 104B catalyst which is about 50% over catalyzation to compensate for excess adsorption of the catalyst by the porous waste.
- the material was heated to 1000° C. in 150 hours to produce a crack-free matrix containing evenly distributed waste material.
- PW-4b is apparently more suitable for containment in vitreous carbon.
- PW-6 is much less, apparently because of its high residual NaNO 3 content, which is about 30 w/o versus almost none in PW-4b.
- the PW-4c-7 is unsuitable because the alumina substrate particles lost their coating of waste and sank to the bottom, finally creating large cracks because of differential shrinkage.
- the particle size of PW-4b is small enough (or the particles are sufficiently friable) that matrix cracking does not occur.
- the method of this invention for encapsulating solid high-level radioactive waste material in vitreous carbon provides an effective and efficient method for preparing these wastes for long-term storage.
Abstract
High-level radioactive wastes are encapsulated in vitreous carbon for long-term storage by mixing the wastes as finely divided solids with a suitable resin, formed into an appropriate shape and cured. The cured resin is carbonized by heating under a vacuum to form vitreous carbon. The vitreous carbon shapes may be further protected for storage by encasement in a canister containing a low melting temperature matrix material such as aluminum to increase impact resistance and improve heat dissipation.
Description
The invention described herein was made in the course of, or under, a contract with the U.S. ENERGY RESEARCH AND DEVELOPMENT ADMINISTRATION.
This invention relates to a method for the preparation of radioactive wastes for storage. More specifically, this invention relates to an improvement in the method for the preparation of high-level radioactive waste solids for storage by encapsulation of the wastes in vitreous carbon.
The chemical reprocessing of spent nuclear reactor fuel elements to recover the unburned nuclear reactor fuel material generates large volumes of aqueous solutions containing radioactive wastes. Not only are there large volumes of solutions, but the solutions are very corrosive and therefore are difficult to handle and store. Since long-term storage of these wastes is necessary to permit decay of the highly radioactive fission products contained therein, the aqueous wastes are converted to a solid form which occupies less volume than the corresponding liquid wastes, is less corrosive and poses less difficult problems in handling and long-term storage. These aqueous radioactive waste solutions can be converted to solid form by spray solidification, fluidized-bed calcination, pot calcination, or by heating to dryness and sintering the resulting solid.
However, the high-level waste solids prepared by the above methods must still be stored for an extended period of time. Many methods for preparing these wastes for storage, so that they pose no present or future threat to the environment, have been and are presently being investigated. For example, the waste might be stored as granular material in metal canisters either in the open, or in water-filled storage tanks. The wastes might be mixed with concrete, asphalt or formed into any one of a number of different mixes of glasses. However, most of these methods have problems attendant with their use for the long-term storage of high-level radioactive wastes. For example, metal canisters can leak, releasing radioactivity to the atmosphere, wastes tied up in concrete or in many glasses may be leached, i.e., water may dissolve soluble radioactive salts which could then contaminate the environment. Glasses may also devitrify which will increase leach rates above acceptable levels and both concrete and glass have poor thermal conductivity which may prevent effective dissipation of the thermal energy generated by the radioactive material.
We have found that, by encapsulating the high-level radioactive solid waste material in vitreous carbon, we are able to eliminate some of the above-enumerated problems. By the method of our invention, powdered high-level radioactive waste solids are mixed with a curable resinous material, formed into an appropriate shape for storage and cured. The cured shape is then heated under a vacuum to 600-1000° C. for a sufficient period of time to carbonize the cured resinous shape to vitreous carbon, encapsulating the high-level radioactive waste solids. Additional physical protection and improved thermal conductivity can be provided the vitreous carbon shape by placing a plurality of the shapes into a metal canister in a regular spaced array and filling the canister including the interstices between the shapes with a low-melting molten matrix material such as aluminum.
The method of the invention has several advantages over the prior art methods for preparing highly radioactive waste solids for storage. For example, the resin and radioactive material can be mixed and the shapes formed at room temperature. The resistance of vitreous carbon to leaching is much better than concrete and generally superior to the glasses, particularly when the glasses are subject to devitrification. Thermal conductivity of the carbon and metal matrix is much better than either concrete, asphalt or glass. In the event that recovery of the wastes is desirable in the future, simple combustion in air of the vitreous carbon will permit easy retrieval.
It is therefore one object of the invention to provide an improved method for preparing high-level radioactive wastes for long-term storage.
It is another object to provide a method for preparing high-level radioactive wastes for long-term storage by encapsulation of the wastes in vitreous carbon.
These and other objects of the invention may be met by mixing the solid high-level radioactive waste material as a -325 mesh powder with a curable resinous material to form a resinous mixture containing about 50 weight percent waste material, casting the resinous mixture into a sheet up to 1/8 inch thick for storage, curing the shaped resinous mixture, and heating the sheet of cured resinous material in a vacuum or inert atmosphere to from 600° to 1000° C. for a period of time sufficient to convert the resinous material to vitreous carbon, thereby providing a sheet of high-level radioactive waste material encapsulated in vitreous carbon ready for long-term storage.
The highly radioactive waste material should be a fine powder no more than about 325 mesh in size for best results. Since there is a large decrease in resin volume during the carbonizing process, larger particle sizes may result in the formation of cracks in the vitreous carbon shapes.
Any solid waste material may be used with the method of the invention as long as it is reasonably inert so that it does not react with the catalyst in the resin. Thus it is preferred that the waste material have been calcined to ensure that only inert oxides are present.
The amount of waste material in the resinous mixture is generally limited to a mixture consistency which can be easily worked. If shapes, such as rods, which can be formed by extrusion are to be prepared, the resinous mixture may contain up to about 70 weight percent waste material, while if flat sheets or spheroids are to be formed, up to about 50 weight percent waste material is the practical limit.
The curable resinous material may be any thermosetting resin which is liquid at room temperature and which has a high carbon yield. One such resin which was found to give good results was polyfurfuryl alcohol although other thermosetting resins such as phenol-formaldehyde can be used in the practice of this invention.
It was generally found that the presence of large quantities of powdered material in the precursor resin required that additional catalyst be added in order to achieve curing within a reasonable period of time. Care must also be taken in order to prevent the curing from proceeding too rapidly and thus producing a large amount of heat which may result in some undesirable cracking.
One limitation of the method of this invention is that in order to prevent the formation of cracks in the shapes during carbonization the shapes are limited to about 1/4 inch in thickness. Three shapes have been found useful with the method of this invention. Resinous mixtures containing up to 70 weight percent solid material may be extruded into rods up to 1/4 inch in cross section for curing, carbonization and subsequent storage. Mixtures containing up to 50 weight percent may be either cast into sheets up to about 1/8 inch in thickness or formed into small spheriods. The spheroids, 200-300 μm in diameter, can be readily prepared by injecting the resinous mixture dropwise into agitated vegetable oil at 100° C. This method is advantageous in that the spheroids are formed and cured within seconds, and could be carbonized in hours.
Carbonization of the cured resinous material containing the solid radioactive waste is readily accomplished by heating the shapes under a vacuum or in an inert atmosphere to from 600° to 1000° C. at a heating rate of about 6° C. per hour to ensure diffusion of gases and complete carbonization of the shapes. Heating should continue for a period of time sufficient to ensure that total carbonization has taken place throughout the entire shape. For rods and sheets of material, carbonization may require from about 100 to 150 hours, while the spheroids 200 to 300 microns in diameter can be carbonized in 9 to 10 hours.
Since the vitreous carbon is brittle and prone to breakage, additional long-term storage protection can be provided by placing a large number of the shapes in a metal canister in a regular spaced array and filling the canister including the interstices between the shapes, with a molten low-melting-temperature matrix material such as aluminum. This matrix material would provide additional impact protection for the vitreous carbon, reduce further any possibility of leaching the radioactive material and further improve the thermal conductivity in the waste container. An advantage of this method is that recovery of the radioactive material still remains a relatively simple procedure, since only the metal need be melted to recover the vitreous carbon sheets.
The following examples are given as illustrating the process of the invention and are not to be taken as limiting the scope or extent of the invention as defined by the appended claims.
To demonstrate the process, Quaker Oats RP 100A resin (polyfurfuryl alcohol) was mixed with about 33 w/o calcined waste with a 400 mesh size. The batch was catalyzed with 8 w/o Quaker Oats RP 104B catalyst which is about 50% over catalyzation to compensate for excess adsorption of the catalyst by the porous waste. Upon completion of curing, the material was heated to 1000° C. in 150 hours to produce a crack-free matrix containing evenly distributed waste material.
To study stability of the resin before conversion to pure carbon, discs of catalyzed polyfurfuryl alcohol resin were cast. The discs were subjected to alpha and gamma radiation, then carbonized to 1000° C. in 150 hours under vacuum. The results are summarized in Table I below. Metallographic examination of carbonized specimens 3 and 7 showed no cracking or evidence of degradation. The gamma dose accumulated by the discs was on the order of 109 rads.
TABLE I
__________________________________________________________________________
Post- Irradiation
% Weight
Appearance
Sample
Irradiation
Irradiation
Weight Gain
Loss on after
No. Treatment
Appearance
(loss), %
Carbonization
Carbonization
__________________________________________________________________________
3 αirradiated
Excellent
0 39.7 Excellent
˜4 × 10.sup.14
particles/cm
from .sup.238 PuO.sub.2
4 None (control)
NA NA 39.8 Excellent
5 γirradiated
Excellent
2%.sup.(a)
40.1 Excellent
2.68 × 10.sup.8 rad
6 γirradiated
Excellent
0.6%.sup.(a)
40.0 Excellent
1.35 × 10.sup.7 rad
7 γirradiated
Excellent
(3.6%).sup.(a)
40.4 Excellent
5.64 × 10.sup.9 rad
__________________________________________________________________________
.sup.(a) Variable weight gain or loss probably because of water
sorption/desorption in gamma facility.
TABLE II
__________________________________________________________________________
Waste Source W/o Waste
W/o Catalyst
Comments
__________________________________________________________________________
Calcined Waste,
50 8 Delayed curing because
Batch PW-4b, -37 micron of catalyst sorption.
(Experiment Standard) No cracking, excellent
microstructure.
PyC - Coated particle
40 4 Curing normal, severely
Batch 5895-85 cracked during
(300 micron) carbonization.
PyC - Coated particle
47 4 Cure normal, completely
Batch 5768-141 disintegrated during
(650 micron) carbonization
__________________________________________________________________________
Polyfurfuryl alcohol resin was mixed with several batches of simulated waste, in all cases adding waste until the mixture could just be poured. The purpose was twofold: (1) to determine how well the vitreous carbon would tolerate different chemical compositions and (2) to see if the vitreous carbon would tolerate a normal particle size distribution, since previous experiments were done with -400 mesh material. Results are summarized in Table III.
TABLE III
__________________________________________________________________________
W/o Waste in Appearance after Carbonization,
Waste Type
Pourable Mix
Curing Behavior
External and Metallographic
__________________________________________________________________________
PW-4b 47 Normal Flat, no cracks, metallographi-
cally very good.
PW-6 36 Normal Slight warping, considerable
cracking, but fairly sound
microstructure.
PW-4c-7
59 Extremely Severely warped and cracked;
slow; required 150° C.
Al.sub.2 O.sub.3 particles sank to bottom,
to harden rest of waste graded according
to settling rate, with almost
none at top (concave surface).
__________________________________________________________________________
PW-4b is apparently more suitable for containment in vitreous carbon. PW-6 is much less, apparently because of its high residual NaNO3 content, which is about 30 w/o versus almost none in PW-4b. The PW-4c-7 is unsuitable because the alumina substrate particles lost their coating of waste and sank to the bottom, finally creating large cracks because of differential shrinkage. Obviously, the particle size of PW-4b is small enough (or the particles are sufficiently friable) that matrix cracking does not occur.
In order to study the resistance of the containment material to leaching water, a number of samples of varying compositions were prepared as described previously and subjected to a standard accelerated leach test. The results of the test are given in Table IV below.
TABLE IV
__________________________________________________________________________
Cumulative
Cumulative
W/o Loss W/o Loss W/o Loss
Sample First 24 hour
Second 24 hour
Third 24 hour
__________________________________________________________________________
Vitreous Carbon
0.09% 0.13% Not leached
No Calcine
82 w/o Vitreous Carbon
0.44% 0.47% 0.52%
18 w/o Calcine
47 w/o Vitreous Carbon
0.27% 0.29% 0.42%
53 w/o Calcine
33 w/o Vitreous Carbon
1.72% 1.86% Not leached
67 w/o Calcine Because >1%
Limit
__________________________________________________________________________
As can be seen from the preceding discussion and examples, the method of this invention for encapsulating solid high-level radioactive waste material in vitreous carbon provides an effective and efficient method for preparing these wastes for long-term storage.
Claims (8)
1. A method of preparing solid high-level radioactive waste material for storage by encapsulating the material in vitreous carbon comprising:
mixing powdered high-level radioactive material less 325 mesh with a curable resinous material selected from the group consisting of polyfurfuryl alcohol and phenolformaldehyde to form a resinous mixture containing up to 70 weight percent waste material;
forming the resinous mixture into appropriate shapes for storage;
curing the shaped resinous mixture; and
heating the cured shapes to from 600° to 1000° C. under a vacuum or inert atmosphere at a rate of about 6° C. per hour for a period of time sufficient to carbonize the cured resinous material to vitreous carbon, thereby forming shapes of high-level radioactive material encapsulated in vitreous carbon for storage.
2. The method of claim 1 wherein the curable resinous material is a polyfurfuryl alcohol.
3. The method of claim 2 wherein the mixture is cured by heating the shapes for about 2 hours at about 70° C.
4. The method of claim 3 wherein the shapes into which the resinous material is formed are selected from the group consisting of rods, sheets and spheroids.
5. The method of claim 4 wherein the shape is a rod up to 1/4 inch in diameter, the rod is formed by extrusion and the resinous mixture contains up to about 70 weight percent waste material.
6. The method of claim 4 wherein the shape is a sheet up to 1/8 inch in thickness, the sheet is formed by casting and the resinous mixture contains up to about 50 weight percent waste material.
7. The method of claim 4 wherein the shape is spheroidal, the spheroids are formed and cured by injecting the resinous mixture dropwise into agitated vegetable oil at 100° C. and the resinous mixture contains up to about 50 weight percent waste material.
8. The method of claim 4 including the additional steps of: placing a plurality of vitreous carbon shapes into a metal canister in a regular spaced array, and filling the canister including the interstices between the carbon shapes with molten aluminum and permitting the aluminum to harden, thereby providing additional protection to the encapsulated high-level waste for long-term storage.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| US05/624,805 US3993579A (en) | 1975-10-22 | 1975-10-22 | Method of encapsulating solid radioactive waste material for storage |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| US05/624,805 US3993579A (en) | 1975-10-22 | 1975-10-22 | Method of encapsulating solid radioactive waste material for storage |
Publications (1)
| Publication Number | Publication Date |
|---|---|
| US3993579A true US3993579A (en) | 1976-11-23 |
Family
ID=24503379
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| US05/624,805 Expired - Lifetime US3993579A (en) | 1975-10-22 | 1975-10-22 | Method of encapsulating solid radioactive waste material for storage |
Country Status (1)
| Country | Link |
|---|---|
| US (1) | US3993579A (en) |
Cited By (8)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US4115311A (en) * | 1977-03-10 | 1978-09-19 | The United States Of America As Represented By The United States Department Of Energy | Nuclear waste storage container with metal matrix |
| EP0057430A1 (en) * | 1981-02-03 | 1982-08-11 | Nukem GmbH | Container for transporting and storing radioactive wastes |
| US4407742A (en) * | 1979-04-28 | 1983-10-04 | Nukem Gmbh | Process for conditioning radioactive and toxic wastes |
| US4496477A (en) * | 1977-04-14 | 1985-01-29 | Ross Donald R | Apparatus and a process for heating a material |
| DE3409803A1 (en) * | 1983-08-18 | 1985-03-07 | Hitachi Zosen Corp., Osaka | Process for the vitrification of radioactive waste |
| US4756869A (en) * | 1983-12-09 | 1988-07-12 | Kernforschungsanlage Julich Gmbh | Method of storing spent nuclear fuel elements |
| US5164123A (en) * | 1988-07-08 | 1992-11-17 | Waste Seal, Inc. | Encapsulation of toxic waste |
| US5169566A (en) * | 1990-05-18 | 1992-12-08 | E. Khashoggi Industries | Engineered cementitious contaminant barriers and their method of manufacture |
Citations (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US3249551A (en) * | 1963-06-03 | 1966-05-03 | David L Neil | Method and product for the disposal of radioactive wastes |
| US3340567A (en) * | 1964-05-05 | 1967-09-12 | Grace W R & Co | Apparatus for producing sol microspheres |
| US3644604A (en) * | 1969-11-14 | 1972-02-22 | Atomic Energy Commission | Nuclear fuel body and process for making same |
| US3854979A (en) * | 1972-06-29 | 1974-12-17 | Aerospace Corp | Process for applying glassy carbon coatings |
-
1975
- 1975-10-22 US US05/624,805 patent/US3993579A/en not_active Expired - Lifetime
Patent Citations (4)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US3249551A (en) * | 1963-06-03 | 1966-05-03 | David L Neil | Method and product for the disposal of radioactive wastes |
| US3340567A (en) * | 1964-05-05 | 1967-09-12 | Grace W R & Co | Apparatus for producing sol microspheres |
| US3644604A (en) * | 1969-11-14 | 1972-02-22 | Atomic Energy Commission | Nuclear fuel body and process for making same |
| US3854979A (en) * | 1972-06-29 | 1974-12-17 | Aerospace Corp | Process for applying glassy carbon coatings |
Non-Patent Citations (1)
| Title |
|---|
| Kaufmann, A. Ed. Nuclear Reactor Fuel Elements Interscience Publishers (a division of John Wiley & Sons), New York, 1962, p. 478. * |
Cited By (8)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| US4115311A (en) * | 1977-03-10 | 1978-09-19 | The United States Of America As Represented By The United States Department Of Energy | Nuclear waste storage container with metal matrix |
| US4496477A (en) * | 1977-04-14 | 1985-01-29 | Ross Donald R | Apparatus and a process for heating a material |
| US4407742A (en) * | 1979-04-28 | 1983-10-04 | Nukem Gmbh | Process for conditioning radioactive and toxic wastes |
| EP0057430A1 (en) * | 1981-02-03 | 1982-08-11 | Nukem GmbH | Container for transporting and storing radioactive wastes |
| DE3409803A1 (en) * | 1983-08-18 | 1985-03-07 | Hitachi Zosen Corp., Osaka | Process for the vitrification of radioactive waste |
| US4756869A (en) * | 1983-12-09 | 1988-07-12 | Kernforschungsanlage Julich Gmbh | Method of storing spent nuclear fuel elements |
| US5164123A (en) * | 1988-07-08 | 1992-11-17 | Waste Seal, Inc. | Encapsulation of toxic waste |
| US5169566A (en) * | 1990-05-18 | 1992-12-08 | E. Khashoggi Industries | Engineered cementitious contaminant barriers and their method of manufacture |
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